Automatically scramming nuclear reactor system

ABSTRACT

An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

FIELD OF THE INVENTION

[0001] The invention generally pertains to nuclear reactor systems, andmore specifically, to automatically scramming nuclear reactor systems.

BACKGROUND OF THE INVENTION

[0002] There are over four-hundred nuclear power plants worldwide,providing nearly twenty percent of the world's electricity. Nuclearpower plants function much like power-generating plants that are fueledby coal or oil. That is, either type of power plant generates heat. Theheat is used to heat water and produce steam, or to heat a gas. Thesteam or the gas, as the case may be, drives one or more turbines whichin turn generate electricity. The difference, of course, is that heat isgenerated at a nuclear power plant by nuclear reactions (i.e., inducedfission) instead of by burning coal or oil.

[0003] Induced fission takes place in the reactor. The fuel for thereactor is provided by a suitable radioactive material (e.g.,uranium-235 or plutonium-239) typically formed into either rods or“pebbles” that are arranged within the core of the reactor. As the fuelfissions, neutrons are released which bombard the nuclei of the otherfuel atoms in the core of the reactor. The bombarded nuclei absorb theneutrons causing the nuclei to become unstable and split, releasing oneor more neutrons which bombard the nuclei of yet other fuel atoms, andso on. The split atoms release energy in the form of radiation and heat.

[0004] During operation of the reactor, a coolant is passed through thecore of the reactor to maintain the reactor at a normal operatingtemperature and keep it from overheating. The coolant may be either agas-phase coolant (e.g., helium) or a liquid-phase coolant (e.g., water)that flows into the reactor, absorbs the heat produced during inducedfission, and flows out of the reactor.

[0005] The heated coolant that flows out of the reactor may then bepassed through a heat-exchanger. Water is also provided to the heatexchanger to absorb heat from the heated coolant. The coolant is thenrecirculated into the reactor. The heat absorbed by the water producessteam. This steam is used to drive the turbines that operate thegenerator and generate electricity. Alternatively, in a direct cyclegas-cooled reactor the cooling fluid is used directly to drive theturbines.

[0006] In some circumstances, the flow of coolant into the reactor maybe insufficient to cool the reactor. As an example, the flow of coolantinto the reactor may be interrupted by a blockage in the pipe system orfailure of a pump, reducing or altogether stopping the flow of coolantinto the reactor. When this happens, the reactor must be shut down sothat the reactor does not overheat.

[0007] The reactor is provided with one or more control elements thatcan be lowered into the reactor to slow and eventually stop thereactions occurring therein when the reactor exceeds a safe operatingtemperature. Control elements may be made from a variety of materialsthat absorb free neutrons. When the control elements are lowered intothe reactor, the control elements absorb the neutrons instead of theneutrons being absorbed by the fuel, causing the reactor to shut down.

[0008] Typically, a number of monitors are used to determine how muchheat is being generated in the reactor. For example, the monitors maymeasure the temperature in the reactor. When the temperature in thereactor exceeds safe operating conditions, the monitors signal anemergency response system which in turn lowers the control elements intothe reactor to shut it down. For safety reasons redundant monitors arecommonly provided so that if one fails, another of the monitors willstill signal the emergency response system of the unsafe operatingcondition so that it can shut down the reactor. However, the monitorsmust still signal the emergency response system when the unsafecondition occurs, thereby introducing delay and another potential pointof failure. In addition, such redundant monitors can be complex andtherefore expensive.

SUMMARY OF THE INVENTION

[0009] An embodiment of an automatically scramming nuclear reactorsystem of the present invention may comprise a core having a coolantinlet end and a coolant outlet end. A cooling system operativelyassociated with the core provides coolant to the coolant inlet end ofthe core and removes heated coolant from the coolant outlet end of thecore. The flow of coolant through the reactor maintains a pressuredifferential between the coolant inlet end of the core and the coolantoutlet end of the core during a normal operating condition of thenuclear reactor system. A guide tube having a first end and a second endis positioned within the core. The first end of the guide tube is influid communication with the coolant inlet end of the core, and thesecond end of the guide tube is in fluid communication with the coolantoutlet end of the core. A control element is positioned within the guidetube and is movable within the guide tube between an upper position anda lower position. The control element automatically falls under theaction of gravity to the lower position when the pressure differentialdrops below a safe pressure differential.

[0010] A method for scramming a nuclear reactor system is also disclosedand may comprise the steps of providing a coolant to a core of thenuclear reactor system at a first pressure, removing heated coolant fromthe core of the nuclear reactor system at a second pressure, the firstpressure being greater than the second pressure during a normaloperating condition of the nuclear reactor system, using a pressuredifferential between the first and second pressures to hold a controlelement above a scramming position during the normal operating conditionof the nuclear reactor system, and the control element automaticallyfalling under the action of gravity to the scramming position when thepressure differential drops below a safe pressure differential.

BRIEF DESCRIPTION OF THE DRAWINGS

[0011] Illustrative and presently preferred embodiments of the inventionare illustrated in the drawings, in which:

[0012]FIG. 1 is an illustration of a nuclear power plant;

[0013]FIG. 2 is a cross-sectional view of one embodiment of anautomatically scramming nuclear reactor system;

[0014]FIG. 3 is a side view of one embodiment of a control element foruse with the automatically scramming nuclear reactor system;

[0015]FIG. 4 is a side view of another embodiment of a control elementfor use with the automatically scramming nuclear reactor system; and

[0016]FIG. 5 is a cross-sectional view of another embodiment of anautomatically scramming nuclear reactor system.

DESCRIPTION OF THE PREFERRED EMBODIMENT

[0017] One embodiments of an automatic scramming system 10 according tothe present invention is shown in FIG. 2 as it may be used in a nuclearpower plant 12 (FIG. 1). Briefly, the nuclear power plant may be ofconventional design (with the exception of the presence of theinvention) and may involve a fission reactor 14 for producing heat. Acoolant 22 is provided to the core 16 of the reactor 14 and absorbs theheat produced during induced fission. The heated coolant 22 flows out ofthe reactor 14 and is passed through a heat exchanger 23. Water 24 isalso provided to the heat exchanger 23 and absorbs the heat from theheated coolant 22 to produce steam 26. The coolant 22 can then berecirculated through the reactor 14, and the steam 26 is used to driveone or more turbines 28, which operate a generator 30 to generateelectricity 32. Of course in another embodiment, the coolant 22 may beused to drive the turbines 28 directly. In any event, if the flow ofcoolant 22 through the reactor 14 is insufficient to cool the core 16 ofthe reactor 14, however, the reactor 14 must be shut down before itoverheats.

[0018] According to the teachings of the present invention, anautomatically scramming nuclear reactor system 10 (FIG. 2) may comprisea core 16 having a coolant inlet end 42 and a coolant outlet end 44. Acooling system 20 (FIG. 1) operatively associated with the core 16 ofthe reactor 14 provides the coolant 22 to the inlet plenum 53 throughthe coolant inlet end 42 of the core 16. As the coolant 22 flows throughthe core 16 of the reactor 14, it absorbs heat. The cooling system 20removes the heated coolant 22 from the outlet plenum 54 through thecoolant outlet end 44 of the core 16, and maintains a pressuredifferential (Δp) between the coolant inlet end 42 of the core 16 andthe coolant outlet end 44 of the core 16 during a normal operatingcondition of the nuclear reactor system 10. A guide tube 50 (FIG. 2)having a first end 51 and a second end 52 is positioned within the core16 so that the first end 51 of the guide tube 50 is in fluidcommunication with the coolant inlet end 42 of the core 16, and thesecond end 52 of the guide tube 50 is in fluid communication with thecoolant outlet 44 end of the core 16. A control element 18 is positionedwithin the guide tube 50 and is movable within the guide tube 50 betweena lower position 19′ and an upper position 19 under the influence of thepressure differential. That is, the control element 18 is raised to theupper position 19 when the flow of coolant 22 is sufficient to cool thecore 16 of the reactor 14 (e.g., a safe pressure differential isestablished) so that induced fission occurs during normal operation. Ifthe flow of coolant 22 is insufficient to maintain the reactor 14 at asafe operating temperature (e.g., the pressure differential drops belowthe safe pressure differential), the control element 18 automaticallyfalls under the action of gravity to the lower position 19′ and shutsdown the reactor 14.

[0019] The automatically scramming nuclear reactor system 10 may beoperated as follows according to the teachings of the invention. Thecooling system 20 provides coolant 22 to the core 16 of the reactor 14at a first pressure (p₁). The heated coolant 22 is then removed from thecore 16 at a second pressure (p₂). The first pressure is greater thanthe second pressure during a normal operating condition of the nuclearreactor system 10. The control element 18 is held above a scrammingposition 19′ (e.g., in position 19) during the normal operatingcondition of the nuclear reactor system 10 by the pressure differential(Δ_(p)) between the first pressure and the second pressure. When thepressure differential drops below a safe pressure differential, thecontrol element 18 automatically falls under the action of gravity tothe scramming position 19′ to shut down the reactor.

[0020] Accordingly, the control elements 18 are raised from the reactor14 under the force of the flow of coolant 22 when it is sufficient tomaintain the reactor 14 at a safe operating condition. A mechanical liftsystem is not required to raise the control elements 18 from the core 16to allow normal operation of the reactor 14. In addition, when the flowof coolant 22 is insufficient to maintain the reactor 14 at a safeoperating temperature, the control elements 18 are automatically loweredinto the core 16 of the reactor 14 under the force of gravity, causingthe reactor 14 to shut down before it overheats. External monitors maybe provided for additional safety, but are not required for operation ofthe automatically scramming nuclear reactor system 10 of the presentinvention.

[0021] Having briefly described one embodiment of an automaticallyscramming nuclear reactor system 10, as well as some of the moresignificant features and advantages thereof, various embodiments of theinvention will now be described in detail.

[0022] A nuclear power plant 12 is illustrated in FIG. 1 in which theautomatically scramming nuclear reactor system 10 (FIG. 2) of thepresent invention may be implemented. According to this embodiment, thenuclear power plant 12 comprises a reactor 14. Fuel 15 is provided in acore 16 of the reactor 14 where induced fission occurs during operation.A cooling system 20 provides coolant 22 (e.g., a gas such as helium or aliquid such as water) through a primary coolant loop (i.e., between acoolant reservoir 21 and the reactor 14). During operation, the coolant22 is pumped from the coolant reservoir 21 into the inlet end 42 of thereactor 14. The coolant is returned from the outlet end 44 of thereactor 14 to a heat exchanger 23 that transfers the heat to water 24 ina secondary coolant loop (i.e., between a water reservoir 25 and theheat exchanger 23). The heated water 24 produces steam 26.

[0023] A steam collection system 27 provides a path for the steam fromthe water reservoir 25 to one or more steam-driven turbines 28. Theturbines 28 are linked to a generator 30 which is operable by therotation of the turbines 28 to generate electricity 32. Of course, in adirect-cycle system having a single loop, the coolant 22 directly drivesone or more of the turbines 28.

[0024] A condenser 34 may be provided to collect the steam 26 from theturbines 28 and convert it to a liquid-phase. A return system (notshown) may provide a path to recirculate the liquid phase into the waterreservoir 25 in a closed loop system. A cooling tower 36 may also beprovided to cool the liquid-phase when it is to be discharged.

[0025] The reactor 14 and the cooling system 20 are preferably containedwithin a housing 40 to reduce the likelihood of radioactive gases orfluids leaking into the surrounding environment and to protect thereactor from external impacts (e.g., by vehicles or airplanes). Thehousing 40 may comprise a concrete liner surrounded by a steelcontainment vessel and an outer concrete building. Of course the housing40 may comprise any suitable barriers based on various designconsiderations. The specific design is typically governed by safety andenvironmental regulations.

[0026] Various ancillary components (not shown) may also be containedwithin the housing 40, such as pumps, electronic controls, monitors,surveillance systems, etc. Such ancillary components are commonlyassociated with nuclear power plants 12, and therefore are not shown ordescribed herein as they are well-understood and further description isnot needed for an understanding of, or to practice the invention.

[0027] The nuclear power plant 12 may be operated to generateelectricity 32 as follows. Radioactive material or fuel 15 (FIG. 2) isprovided in the core 16 of the reactor 14 where it undergoes inducedfission and releases heat. During the reaction, the coolant 22 flowsthrough the core 16 of the reactor 14 to absorb heat from the reactionand maintain the reactor 14 at a normal operating temperature so that itdoes not overheat. The coolant 22 flows into the inlet end 42 of thereactor 14, absorbs heat in the core 16, and flows out through theoutlet end 44 of the reactor 14. The heat absorbed by the coolant 22 istransferred to the water 24 by the heat-exchanger 23. The coolant isthen recirculated into the reactor 14, and the heat absorbed by thewater 24 produces steam 26.

[0028] The steam 26 is collected by the steam collection system 27 andused to drive the turbines 28, which in turn operate the generator 30 togenerate electricity 32. The steam 26 is then collected from theturbines 28 by the condenser 34 and converted to a liquid phase. Theliquid phase may be returned to the water reservoir 25 to enhance theefficiency of the nuclear power plant 12. Alternatively, the liquidphase may be passed through the cooling tower 36 and then dischargedinto the environment (e.g., as indicated by arrow 38).

[0029] The foregoing description of the nuclear power plant 12 isprovided in order to better understand one environment in which thevarious embodiments of the automatic scramming system 10 of the presentinvention may be used. However, it should be understood that theautomatic scramming system 10 of the present invention may be used inany of a wide range of other types of nuclear power plants 12 now knownor that may be developed in the future. Consequently, the automaticscramming system 10 of the present invention should not be regarded asbeing limited to use with the nuclear power plant 12 shown and describedherein. Also, since a more detailed description of the nuclear powerplant 12 is not required to understand or practice the invention, thenuclear power plant 12 that the automatic scramming system 10 of thepresent invention may be used in conjunction with will not be describedin further detail herein.

[0030] The reactor 14 is shown in more detail in FIG. 2. The reactor 14may comprise a vessel 13 surrounding a core 16. The cooling system 20provides coolant into an inlet plenum 53 of the core 16 through acoolant inlet end 42 and removes heated coolant from an outlet plenum 54of the core 16 through a coolant outlet end 44 of the reactor 14. Aspreviously discussed, the fuel 15 may comprise any suitable radioactivematerial and may be formed into “pebbles” and provided in the form of a“bed” within the core 16 of the reactor 14. Such a reactor is commonlyreferred to as a pebble-bed reactor.

[0031] For purposes of illustration, the reactor 14 may be patterned onthe South African utility Eskom pebble-bed modular reactor (PBMR).Exemplary design parameters for such a reactor are given in Table 1.TABLE 1 Core height 10 meters (m) Core diameter 3 m Fuel UO₂ Reflectormaterial graphite Reflector thickness 1 m (all around) Fuel packingfraction in core 0.61

[0032] Also according to one exemplary embodiment, the fuel 15 maycomprise TRISO-coated uranium oxide (UO₂) microspheres embedded in aspherical graphite matrix inside a shell of pure graphite. The “pebbles”15 in this example are packed in the core 16 of the reactor 14 with apacking fraction of about 0.61, although this may vary in other designs.

[0033] It is noted that the fuel concentration may be adjusted toproduce a critical core 16 when the control elements 18 are suspendedabove the reactor 14. It is further noted that the reactor 14 describedin Table 1 is merely exemplary of one reactor that may utilize theautomatic scramming system 10 of the present invention. Indeed, theinvention is not limited to use with a pebble bed reactor and may beused with any suitable reactor, now known or that may be developed inthe future. For example, the invention may be used with a prismaticcore.

[0034] The cooling system 20 provides the coolant 22 into the inletplenum 53 of the core 16 of the reactor 14 through an inlet end 42. Forexample, the coolant 22 may be provided into the reactor 14 in adownward flow, as illustrated by arrow 45 in FIG. 2. The coolant 22flows through the core 16 of the reactor 14 in a downward direction asillustrated by arrows 41 and is exhausted from the outlet plenum 54through an outlet 44 in the direction of arrow 46, as shown in FIG. 2.

[0035] One or more guide tubes 50 may also be provided in the vessel 13of the reactor 14 in any suitable area of neutronic importance. In oneembodiment, the guide tube 50 is positioned within the core 16 as shownin FIG. 2. However, other arrangements are also possible, where one ormore of the guide tubes 50 are provided adjacent the core 16 (e.g., inthe reflector region). According to one embodiment of the automaticscramming system 10, the guide tube 50 has a first end 51 and a secondend 52. The first end 51 of the guide tube 50 is in fluid communicationwith the inlet plenum 53 (e.g., the inlet end 42 of the core 16), andthe second end 52 of the guide tube 50 is in fluid communication withthe outlet plenum 54 (e.g., the outlet end 44 of the core 16).

[0036] The guide tube 50 may be any suitable shape. In one embodiment,the guide tube 50 comprises one or more bends and is generally shaped asa “trombone” or a “paperclip”, the reason for which will become apparentin the following discussion. In addition, the guide tube 50 may be madeof any suitable material. In one embodiment, the guide tube 50 may bemanufactured of stainless steel. However, other embodiments are alsocontemplated as being within the scope of the invention. Indeed, theparticular design of the guide tube 50 may vary based on designconsiderations and is not to be limited to the particular embodimentshown and described here.

[0037] A control element 18 is positioned within each of the one or moreguide tubes 50. The control element 18 is made of any suitable materialthat absorbs neutrons. Hence, when the control element 18 is loweredinto the core 16 of the reactor 14, the control element 18 absorbsneutrons and slows the rate of reaction, eventually causing the reactor14 to shut down.

[0038] According to one embodiment of the automatic scramming system 10,the control element 18 is movable within the guide tube 50 between anupper position 19 and a lower position 19′. When the control element 18is raised into the upper position 19, the reaction proceeds therein.However, when the control element 18 is lowered into the lower position19′, the control element slows the rate of the reaction occurringtherein and eventually shuts down the reactor 14.

[0039] The particular design of the control element 18 may varyaccording to the teachings of the invention as discussed in furtherdetail below. Design parameters for one embodiment of the controlelement 18 are given in TABLE 2. TABLE 2 Design Parameter Value Controlelement material boron carbide Control element length 1 m or 2 m Controlelement diameter 1 cm or 2.5 cm Control element cladding materialstainless steel Control element cladding thickness 1 mm

[0040] Of course the automatic scramming system 10 of the presentinvention may comprise any suitable number of control elements 18. Inone embodiment there are four control elements, arranged in a circlehaving a radius of 75 cm. However, the particular configuration andarrangement of the control rod(s) 18 in the core 16 of the reactor 14will depend on various design considerations, as will become apparent inthe detailed discussion that follows.

[0041] According to one embodiment of the invention, the control element18 may be a piston body 18′ closely received within the guide tube 50,such as shown in FIG. 3. Little, if any coolant 22 flows around thepiston body 18′. Where desired, any leakage around the piston body 18′may be further minimized by loose seals (not shown) provided between thepiston body 18′ and the walls 56 of the guide tube 50. Accordingly, thepiston body 18′ has a relatively high lift capacity because nearly allof the differential pressure between the inlet and outlet plena 53, 54is available to raise it out of the core 16 of the reactor 14 (e.g.,into position 19), as illustrated by arrows 60 in FIG. 3. When thepressure differential is less than the weight of the piston body 18′,considering the influence of friction, which is minimal, the piston body18′ drops into the core 16 of the reactor 14 under the force of gravity(e.g., into position 19′).

[0042] According to another embodiment of the invention, the controlelement 18 may be an aerodynamic body 18″ such as shown in FIG. 4. Theaerodynamic body 18″ is loosely received within the guide tube 50 and islifted from the core 16 of the reactor 14 (e.g., into position 19)primarily by a dynamic pressure component of the bypass flow of coolantpast the aerodynamic body. That is, the aerodynamic body 18″ is liftedprimarily by the drag force of the coolant moving through the guide tube50 past the aerodynamic body 18″, as illustrated by arrows 61 in FIG. 4.Again, the aerodynamic body 18″ is lowered into the core 16 of thereactor 14 (e.g., into position 19′) when the flow of coolant 22decreases sufficiently for the weight of the control element to exceedthe drag force of the flow of coolant 22.

[0043] The aerodynamic body 18″ may be loosely received within the guidetube 50 with any suitable clearance therebetween. The particular designwill be based on design considerations, such as will become apparent inthe following discussion. As an example, however, the clearance betweenthe perimeter of the piston body 18′ and the inside of the guide tube 50may be in the range of about 0.5 mm and 1.0 mm. The clearance betweenthe perimeter of the aerodynamic body 18″ and the inside of the guidetube 50 may be on the order of about 10%±1%. In one embodiment, theclearance is about 3.0 mm. Again, it is noted that these approximationsare only provided for illustrative purposes.

[0044] Either embodiment of the control element 18 may be used with theautomatic scramming system 10 according to the teachings of theinvention. Although the piston body 18′ may have a greater lift capacitythan the aerodynamic body 18″, the aerodynamic body 18″ may drop morereadily under the force of gravity when the coolant flow decreases,allowing for faster shut down of the reactor 14. In addition, theaerodynamic body 18″ and has a lower probability of binding within theguide tube 50 than the piston body 18′.

[0045] It is noted that other suitable designs of the control element 18are also possible and will become apparent to one skilled in the artafter having become familiar with the teachings of the presentinvention. For example, the control element 18 may be shaped as a ball,a collection of balls, a cylinder, etc. As another example, theaerodynamic body 18″ may be provided with fins or shallow channels onthe side. The fins or shallow channels may be designed to cause thecontrol element 18 to slowly spin, reducing the tendency to vibrate inthe guide tube 50.

[0046] The control element(s) 18 are designed with sufficient shutdownreactivity. That is, the control elements 18 are designed to absorbenough neutrons when lowered into the core 16 that the reaction slowsand eventually stops, shutting down the reactor 14. The change inreactivity resulting from insertion of the control elements 18 into thecore may be modeled, as shown in Table 3 for three combinations ofcontrol element length and diameter. TABLE 3 Design Design Design 1 2 3Control element length (m) 1.0 1.0 2.0 Control element diameter (cm) 1.02.5 2.5 Control element mass (kg) 0.244 1.82 3.64 k_(eff) (withdrawn)1.00441 0.98793 0.98793 k_(eff) (inserted) 0.99940 0.97252 0.96452Reactivity (4 elements) ($) 0.244 1.8 3.6

[0047] The results presented in Table 3 indicate that Design 1 may beinsufficient for a secure reactor scram, but may be sufficient tomaintain the reactor in a shutdown state. After the core 16 of thereactor 14 cools following shutdown, the inserted control elements 18would prevent re-criticality. Even Design 2 provides marginal scramreactivity. However, Design 3 provides ample shutdown reactivity withcontrol elements 18 that can be supported by the available lift force.Of course it is understood the values presented in Table 3 are merelyexemplary of the shutdown reactivity of the control elements 18, andthat the shutdown reactivity of the control elements 18 may bedetermined for any of a variety of different designs of the controlelements 18.

[0048] Operation of the automatic scramming system 10 is as follows. Thecooling system 20 provides coolant 22 to the inlet plenum 53 through theinlet end 42 of the core 16 and removes heated coolant 22 from theoutlet plenum 54 through the outlet end 44 of the core 16. Accordingly,the flow of coolant 22 through the reactor 14 maintains a pressuredifferential between the coolant inlet end 42 of the core 16 and thecoolant outlet end 44 of the core 16 (i.e., between the inlet and outletplena 53, 54) during a normal operating condition of the nuclear reactorsystem 10. This pressure differential may be used to raise and lower thecontrol element(s) 18 in the guide tube 50.

[0049] More specifically, the coolant 22 enters the pebble bed in thecore 16 of the reactor 14 from the inlet plenum 53, then the coolant 22accelerates and turns numerous times as it moves through the packedpebble bed, and finally slows down as it moves into the outlet plenum54. As such, the flow of coolant 22 through the pebble bed sustains apressure drop (Δp) between the pressure in the inlet plenum (p₁) and thepressure in the outlet plenum (p₂) during normal operation. For heliumin a reactor 14 such as previously described, this pressure drop is inthe range of about 110 to 203 kilo-Pascals (kPa). The differentialpressure in the core 16 of the reactor 14 causes the control element 18to be lifted as a fraction of the coolant 22 enters the open end 51 ofthe guide tube 50 in the inlet plenum 53.

[0050] Also according to the teachings of the invention, the controlelement 18 automatically falls under the action of gravity to the lowerposition 19′ when the pressure differential drops below a safe pressuredifferential. For example, the control element 18 falls into the core 16of the reactor 14 (e.g., position 19′) when the pressure within the core16 drops below that necessary for effective cooling of the reactor 14.Similarly, the control element 18 falls into the core 16 of the reactor14 when the flow rate of the coolant 22 drops below that necessary foreffective cooling of the reactor 14.

[0051] The control element 18 drops if the core 16 experiences a loss offorced cooling (either with or without depressurization). For example,where the cooling blowers fail or are turned off, or the coolant pathbecomes obstructed. In such an event, the pressure differential betweenthe inlet and outlet plena 53, 54 is negligible and the control element18 drops into the core 16 of the reactor 14. As the core 16depressurizes, the control element 18 drops into a portion of the guidetube 50 that is surrounded by reactor core (e.g., position 19′) as soonas the flow of coolant 22 decreases below a limiting or minimum value atwhich the lifting force balances the gravitational force.

[0052] The force (F_(L)) required to lift the control element 18 mustexceed the gravitational force (F_(g)) to raise it out of the core 16 ofthe reactor 14. The available lift for the piston body 18′ is a functionof the projected surface area (i.e., the static pressure component), asnearly the entire pressure differential is available to act on thecontrol element 18′. As an example, a diameter of 2.5 cm is sufficientto lift and support a 10 kg piston configuration control element with apressure differential of 203 kPa.

[0053] The following illustrates calculations that can be used tooptimize the design of an aerodynamic body 18″ so that the lifting forceexceeds the gravitational force under normal operating conditions.First, the drag force (F_(d)) required to lift the aerodynamic body 18″(e.g., from position 19′ to position 19) may be determined as follows:$\begin{matrix}{F_{d} = {\frac{1}{2}C_{d}A_{x}{\rho\upsilon}^{2}}} & (1)\end{matrix}$

[0054] where:

[0055] ν is the gas velocity past the control element;

[0056] ρ is the coolant density at flow conditions;

[0057] A_(x) is the cross section area of the rod; and

[0058] C_(d) is the drag coefficient.

[0059] In the above equation (1), ρν² is the dynamic pressure component.Equation (1) may be solved for the velocity as follows: $\begin{matrix}{v = \sqrt{\frac{2F_{d}}{\rho \quad C_{d}A_{x}}}} & (2)\end{matrix}$

[0060] Accordingly, equation (2) can be solved to determine the gasvelocity that is required past the aerodynamic body 18″.

[0061] The coolant pressure drop due to friction (Δp_(f)) in the guidetube 50 can be determined as follows: $\begin{matrix}{{\Delta \quad p_{f}} = {\frac{1}{2}{{\rho\upsilon}^{2}\left\lbrack {{f\frac{L}{D_{ann}}} + K} \right\rbrack}}} & (3)\end{matrix}$

[0062] where:

[0063] ƒ is the Moody friction factor;

[0064] L is the length of the control element;

[0065] D_(ann) is the effective diameter of the annular gap between thecontrol element 18″ and the guide tube 50; and

[0066] K is the sum of the entrance and exit losses.

[0067] Thus, the available drag force can be estimated as follows. Usingequation (3), the flow velocity in the gap around the control elementcan be expressed as a function of the total available pressure dropacross the core (Δp_(c)) using equation (1) and replacing Δp_(f) withΔp_(c). The maximum available drag force is obtained by using flowvelocity in equation (1) as follows: $\begin{matrix}{F_{d} = {\frac{C_{d}A_{x}}{\left\lbrack {{f\frac{L}{D_{ann}}} + K} \right\rbrack}\Delta \quad p_{c}}} & (4)\end{matrix}$

[0068] Acceleration or deceleration along the guide tube 50 due toheating or cooling in the lower and upper flow regions of the guide tube50, respectively, may be ignored as they roughly cancel one another.Therefore, equation (4) may be used to determine the maximum dragavailable under normal operating conditions.

[0069] In some circumstances, it may be desirable to operate the reactor14 at reduced power. When the reactor 14 is operated at a reduced power,the flow rate of the coolant 22 through the core 16 is decreased tomaintain an optimum operating temperature. Of course, decreasing thecoolant flow also decreases the available pressure differential acrossthe core 16 of the reactor, and therefore reduces the lifting forceavailable to raise the aerodynamic body 18″.

[0070] The minimum drag force to lift a given aerodynamic body 18″ canbe determined as follows. Using equation (3) and the requirement thatessentially all the power generated in the core 16 is carried away bythe flow of coolant 22, the frictional pressure drop may be expressed asa function of the power (P) as follows:

P=mc _(p) ΔT  (5)

[0071] where:

[0072] c_(p) is the heat capacity of the coolant; and

[0073] ΔT is the temperature change of the coolant as it traverses thecore; and

[0074] m is the mass flow rate m.

[0075] The mass flow rate (m) may be determined as follows:

m=ρν  (6)

[0076] where:

[0077] A is the total effective flow path cross section area.

[0078] Therefore, the flow velocity (v) can be expressed as a functionof the power (P). That expression is then squared and used to replace v²in equation (3) as follows: $\begin{matrix}{{\Delta \quad p_{f}} = {{\frac{1}{2}\left\lbrack \frac{{f\frac{L}{D_{ann}}} + K}{\rho \quad A^{2}{c_{p}^{2}\left( {\Delta \quad T} \right)}^{2}} \right\rbrack}P^{2}}} & (7)\end{matrix}$

[0079] When the reactor 14 is operated at a power less than a limitingvalue, the control element 18″ drops into the core 16 and shuts thereactor 16. An estimate of the limiting power can be determined asfollows. The minimum coolant velocity to maintain the control element18″ out of the core can be determined from equation (2) with the dragforce equated to the weight of the rod. The drag pressure drop can becalculated using equation (3). Finally, using equation (7), it isdetermined that this pressure drop corresponds to a percentage of thefull operating power. When the power drops below this percentage of fullpower, the control elements lower into the core 16 of the reactor toshut it down.

[0080] The aerodynamic control element 18″ may also be designed based onthe time of descent into the core 16 of the reactor 14. To simplify thisdiscussion, assume that the control element 18″ remains suspended in theraised position 19 until depressurization is complete, and then itstarts to drop into the lower position 19′. The following equation canthen be used to determine the time of descent of the control element 18″into the core 16. $\begin{matrix}{{M\frac{^{2}x}{t^{2}}} = {{Mg} - {\frac{1}{2}C_{d}A_{x}\rho \quad v^{2}}}} & (8)\end{matrix}$

[0081] where:

[0082] M is the mass of the control element;

[0083] x is position; and

[0084] t is time.

[0085] Preferably, the aerodynamic control element 18″ is designed sothat it drops well in advance of what is 30 required to hold down thereactivity in the core 16 of the reactor 14.

[0086] It is readily apparent that there are various designconsiderations that will affect the performance of the automaticallyscramming nuclear reactor system 10 of the present invention. Thesedesign considerations may include the position of the guide tube(s) 50in the core 16 of the reactor 14 to target areas of greater neutronicimportance. Other design considerations may include any of a number offactors that may be varied to optimize the configuration of the controlelement(s) 18 and/or guide tube(s) 50 to provide more or less lift. Forexample, the lift can be increased considerably with only small changesin the diameter of the control element 18. A larger cross sectionprovides more surface area for the lift force. A larger cross sectionmay allow operation at a lower fraction of the maximum power, but thecontrol element 18 may weigh more. Of course in another embodiment, thecontrol element 18 may be made with a hollow central region so as toweigh less. Yet other design parameters will become readily apparent toone skilled in the art after having become familiar with the teachingsof the present invention.

[0087] An alternative embodiment of the automatic scramming nuclearreactor system 110 is shown in FIG. 5. According to this embodiment, thecoolant 22 may be provided into the reactor 114 in an upward flow. Thatis, the coolant 22 is provided into the inlet plenum 153 through theinlet end 142 of the reactor 114 in the direction of arrow 145. Thecoolant 22 flows through the core 116 of the reactor 114 in an upwarddirection as illustrated by arrows 141. The heated coolant 22 is thenexhausted from the outlet plenum 154 through an outlet end 144 of thereactor 114 in the direction of arrow 146.

[0088] According to this embodiment, the guide tube 150 is positioned inthe core 116 of the reactor 114 so that it is open on one end 151 to theinlet plenum 153 and on the other end 152 to the outlet plenum 154. Thepressure differential established between the inlet and outlet plena153, 154 causes the control rod 118 to rise in the guide tube 150 (e.g.,to position 119) during a normal operating condition. When the flow isinsufficient to cool the core 116 of the reactor 114, the control rod118 drops under the influence of gravity into the core 116 (e.g., toposition 119′) and shuts down the reactor 114.

[0089] Of course, it is understood that other embodiments of theautomatic scramming nuclear reactor system 110 are also possible. Forexample, either the piston body 18′ or the aerodynamic body 18″described above may be used with the automatic scramming nuclear reactorsystem 110 of the present invention.

[0090] It is readily apparent that according to embodiments of theinvention the automatic scramming system 10 responds to changes in theflow of coolant 22 to raise and lower the control elements 18 in thereactor 14. Furthermore, the control elements 18 are lowered into thereactor 14 under the force of gravity to automatically shut down thereactor 14 before it can overheat. Consequently, the claimed inventionrepresents an important development in the field of nuclear powergeneration.

[0091] Having herein set forth preferred embodiments of the presentinvention, it is anticipated that suitable modifications can be madethereto which will nonetheless remain within the scope of the presentinvention. Therefore, it is intended that the appended claims beconstrued to include alternative embodiments of the invention exceptinsofar as limited by the prior art.

What is claimed is:
 1. An automatically scramming nuclear reactorsystem, comprising: a core, said core having a coolant inlet end and acoolant outlet end; a cooling system operatively associated with saidcore, said cooling system providing coolant to the coolant inlet end ofsaid core and removing heated coolant from the coolant outlet end ofsaid core, said cooling system maintaining a pressure differentialbetween the coolant inlet end of said core and the coolant outlet end ofsaid core during a normal operating condition of said nuclear reactorsystem; a guide tube positioned within said core, said guide tube havinga first end and a second end, the first end of said guide tube being influid communication with the coolant inlet end of said core, the secondend of said guide tube being in fluid communication with the coolantoutlet end of said core; and a control element positioned within saidguide tube, said control element being movable within said guide tubebetween an upper position and a lower position, the control elementautomatically falling under the action of gravity to the lower positionwhen the pressure differential drops below a safe pressure differential.2. The nuclear reactor system of claim 1, wherein said guide tubecomprises at least one straight section between the first and secondends, said control element being moveable within the straight sectionbetween the upper and lower positions.
 3. The nuclear reactor system ofclaim 1, wherein said coolant comprises a gas.
 4. The nuclear reactorsystem of claim 1, wherein said core is selected from the groupconsisting of a pebble bed, and a prismatic core.
 5. The nuclear reactorsystem of claim 1, wherein said control element comprises a piston body,said piston body being closely received within said guide tube, saidpiston body being held above the lower position within said guide tubeprimarily by a static pressure component of the pressure differential.6. The nuclear reactor system of claim 1, wherein said control elementcomprises an aerodynamic body, said aerodynamic body being looselyreceived within said guide tube, said aerodynamic body being liftedabove the lower position within said guide tube primarily by a dynamicpressure component of a bypass flow of coolant in said guide tube.
 7. Anautomatically scramming nuclear reactor system, comprising: a core, saidcore having a coolant inlet end and a coolant outlet end; a coolingsystem operatively associated with said core, said cooling systemproviding coolant to the coolant inlet end of said core and removingheated coolant from the coolant outlet end of said core, said coolingsystem maintaining a pressure differential between the coolant inlet endof said core and the coolant outlet end of said core during a normaloperating condition of said nuclear reactor system; a guide tubepositioned adjacent said core, said guide tube having a first end and asecond end, the first end of said guide tube being in fluidcommunication with the coolant inlet end of said core, the second end ofsaid guide tube being in fluid communication with the coolant outlet endof said core; and a control element positioned within said guide tube,said control element being movable within said guide tube between anupper position and a lower position, the control element automaticallyfalling under the action of gravity to the lower position when thepressure differential drops below a safe pressure differential.
 8. Thenuclear reactor system of claim 7, wherein said control elementcomprises a piston body, said piston body being closely received withinsaid guide tube, said piston body being held above the lower positionwithin said guide tube primarily by a static pressure component of thepressure differential.
 9. The nuclear reactor system of claim 7, whereinsaid control element comprises an aerodynamic body, said aerodynamicbody being loosely received within said guide tube, said aerodynamicbody being lifted above the lower position within said guide tubeprimarily by a dynamic pressure component of a bypass flow of coolant insaid guide tube.
 10. A method for scramming a nuclear reactor system,comprising: providing a coolant to a core of said nuclear reactorsystem; removing heated coolant from the core of said nuclear reactorsystem, providing coolant and removing heating coolant establishing aflow of coolant through said nuclear reactor system, the flow of saidcoolant providing a lift force during a normal operating condition ofsaid nuclear reactor system; using the lift force provided by the flowof said coolant to hold a control element above a scramming positionduring the normal operating condition of said nuclear reactor system,the control element automatically falling under the action of gravity tothe scramming position when said nuclear reactor system exceeds a safeoperating temperature.
 11. The method of claim 10, wherein using thelift force to hold a control element above a scramming positioncomprises using primarily a static pressure component of the pressuredifferential to hold the control element above the scramming position.12. The method of claim 10, wherein using the lift force to hold acontrol element above a scramming position comprises using primarily adynamic pressure component of a bypass flow of coolant to hold thecontrol element above the scramming position.
 13. The method of claim10, wherein said step of providing a coolant comprises providing acoolant in a gas phase.
 14. An automatically scramming nuclear reactorsystem, comprising: a core; cooling means operatively associated withsaid core for providing coolant to said core, for removing heatedcoolant from said core, and for maintaining a pressure differentialbetween the coolant provided to said core and the heated coolant removedfrom said core during a normal operating condition of said nuclearreactor system; and control element means for scramming said core whensaid control element means is lowered to a scramming position, thepressure differential holding said control element means above thescramming position during the normal operating condition of said nuclearreactor system, said control element means automatically falling underthe action of gravity to the scramming position when the pressuredifferential drops below a safe pressure differential.
 15. Anautomatically scramming nuclear reactor system, comprising: a vesseldefining an inlet plenum and an outlet plenum; a cooling systemoperatively associated with said nuclear reactor system, said coolingsystem providing a coolant to the inlet plenum and removing heatedcoolant from the outlet plenum, said cooling system maintaining apressure differential between the inlet plenum and the outlet plenumduring a normal operating condition of said nuclear reactor system; aguide tube positioned within said vessel, said guide tube having a firstend and a second end, the first end of said guide tube being in fluidcommunication with the inlet plenum, the second end of said guide tubebeing in fluid communication with the outlet plenum; and a controlelement positioned within said control element guide tube, said controlelement being movable within said guide tube between an upper positionand a lower position, said control element automatically falling underthe action of gravity to the lower position when the pressuredifferential drops below a safe pressure differential.
 16. The nuclearreactor system of claim 15, wherein said guide tube comprises at leastone straight section between the first and second ends, said controlelement being moveable within the straight section between the upper andlower positions.
 17. The nuclear reactor system of claim 15, whereinsaid coolant comprises a gas.
 18. The nuclear reactor system of claim15, wherein said vessel comprises a core having a pebble bed.
 19. Thenuclear reactor system of claim 15, wherein said control element isclosely received within said guide tube, said control element being heldabove the lower position within said guide tube primarily by a staticpressure component of the pressure differential.
 20. The nuclear reactorsystem of claim 15, wherein said control element is loosely receivedwithin said guide tube, said control element being lifted above thelower position within said guide tube primarily by a dynamic pressurecomponent of a bypass flow of coolant in said guide tube.
 21. Anautomatically scramming nuclear reactor system, comprising: a core;cooling means operatively associated with said core for providingcoolant to an inlet plenum adjacent said core, for removing heatedcoolant from an outlet plenum adjacent said core, and for maintaining apressure differential between the inlet plenum and the outlet plenumduring a normal operating condition of said nuclear reactor system;control element means for scramming said core when said control elementmeans is lowered to a scramming position guide tube means positionedwithin said core and fluidically connected to the inlet plenum and saidoutlet plenum for receiving said control element means, for allowingsaid control element means to be held above the scramming position bythe pressure differential between the inlet plenum and the outlet plenumduring the normal operating condition, and for allowing said controlelement means to fall under the action of gravity to the scrammingposition when the pressure differential drops below a safe pressuredifferential.